Subchannel analysis of a small ultra-long cycle fast reactor core

2014 ◽  
Vol 270 ◽  
pp. 389-395 ◽  
Author(s):  
Han Seo ◽  
Ji Hyun Kim ◽  
In Cheol Bang
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Long Cycle ◽  
Subchannel Analysis

Development of the LMFBR Subchannel Analysis Code ATHAS-LMR and Analysis of Blockage Accident

2018 ◽  
Author(s):  
Peng Du ◽  
Jianqiang Shan ◽  
Bo Zhang
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
High Flow ◽  
Analysis Program ◽  
Pressurized Water Reactor ◽  
Flow Area ◽  
Pressurized Water ◽  
Original Grid ◽  
Flow Blockage ◽  
Subchannel Analysis

ATHAS is the subchannel analysis code, independently developed by Nuclear Safety and Operation Laboratory (NUSOL) of Xi'an Jiao Tong University (XJTU), the original version of the ATHAS is only suitable for pressurized water reactor. This paper is aimed at the characteristics of sodium cooled fast reactor core mentioned above, developing the analysis program AYHAS-LMR, which is applicable for LMFBR. In the design of sodium cold fast reactor, flow blockage must be considered to meet the design criteria. In view of this, a new model is proposed in this paper and added in ATHAS-LMR, considering the flow blockage which changes the flow area and the wetted perimeter. This model is based on the original grid model, we use the grid to simulate the presence of block, increase the resistance of the blockage area artificially. Then, ATHAS-LMR is able to simulate the flow blockage in LMFBR. In this paper, simulations of FFM5B blockage test were done for the high flow case with ATHAS-LMR, and we compare the simulation results with COMMIX-1. The result indicates that ATHAS-LMR is suitable for the analysis of the flow blockage.

Power Flattening Study of Ultra-Long Cycle Fast Reactor Core

2019 ◽  
pp. 373-380
Author(s):  
Taewoo Tak ◽  
Jiwon Choe ◽  
Yongjin Jeong ◽  
Jinsu Park ◽  
Deokjung Lee ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Long Cycle ◽  
Power Flattening

Optimization study of Ultra-long Cycle Fast Reactor core concept

2014 ◽  
Vol 73 ◽  
pp. 145-161 ◽  
Author(s):  
Taewoo Tak ◽  
Deokjung Lee ◽  
T.K. Kim ◽  
Ser Gi Hong
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Core Concept ◽  
Long Cycle ◽  
Optimization Study

Advances in the analysis of fast reactor core and coolant circuit structures

1992 ◽  
Vol 134 (1) ◽  
pp. 37-58
Author(s):  
Y.W. Chang ◽  
D.T. Eggen ◽  
A. Imazu ◽  
M. Livolant
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Coolant Circuit

Concept of hydride fuel target subassemblies in a fast reactor core for effective transmutation of MA

1998 ◽  
Vol 271-273 ◽  
pp. 530-533 ◽  
Author(s):  
M Yamawaki ◽  
H Suwarno ◽  
T Yamamoto ◽  
T Sanda ◽  
K Fujimura ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Reactor Core

Calculation studies of coated particles performance in sodium-cooled fast reactor

Kerntechnik ◽  
2021 ◽  
Vol 86 (1) ◽  
pp. 45-49
Author(s):  
N. V. Maslov ◽  
E. I. Grishanin ◽  
P. N. Alekseev
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Heat Removal ◽  
Design Solution ◽  
Fast Neutrons ◽  
Steel Shell ◽  
Primary Circuit ◽  
Coated Particles ◽  
The Core ◽  
Fuel Assemblies

Abstract This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

CFD Investigation of Thermal-Hydraulic Behaviors in Full Reactor Core for Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Jing Chen ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Kui Zhang ◽  
Mingjun Wang ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Power Distribution ◽  
Three Dimensional ◽  
Reactor Core ◽  
Heat Removal ◽  
The Core ◽  
Computational Fluid Dynamics Cfd ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Full Core

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.

scholarly journals Burnup concept for a long-life fast reactor core using MCNPX.

2013 ◽  
Author(s):  
Holschuh, Thomas Vernon, ◽  
Lewis, Tom Goslee, ◽  
Parma, Edward J.,
Keyword(s):  
Fast Reactor ◽  
Reactor Core ◽  
Long Life
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